V. Soukhanovskii
A5051599621- Magnetic confinement fusion research
- Fusion materials and technologies
- Superconducting Materials and Applications
- Particle accelerators and beam dynamics
- Ionosphere and magnetosphere dynamics
- Plasma Diagnostics and Applications
- Nuclear reactor physics and engineering
- Laser-Plasma Interactions and Diagnostics
- Solar and Space Plasma Dynamics
- Atomic and Molecular Physics
- Nuclear Physics and Applications
- Laser-induced spectroscopy and plasma
- Nuclear Materials and Properties
- Korean Peninsula Historical and Political Studies
- Semiconductor materials and devices
- Spacecraft and Cryogenic Technologies
- Muon and positron interactions and applications
- Nuclear Engineering Thermal-Hydraulics
- Particle Accelerators and Free-Electron Lasers
- Atomic and Subatomic Physics Research
- Parallel Computing and Optimization Techniques
- Metal and Thin Film Mechanics
- Advanced Data Storage Technologies
- Spectroscopy and Laser Applications
- Catalytic Processes in Materials Science
Lawrence Livermore National Laboratory
2015-2024
Princeton Plasma Physics Laboratory
2001-2016
Oak Ridge National Laboratory
2003-2012
University of California, San Diego
2006-2009
Lawrence Livermore National Security
2005-2009
Johns Hopkins University
1996-2006
Princeton University
2002-2006
Higashihiroshima Medical Center
2006
Hiroshima University
2006
University of California, Irvine
2006
Abstract The objectives of NSTX-U research are to reinforce the advantages STs while addressing challenges. To extend confinement physics low- A , high beta plasmas lower collisionality levels, understanding transport mechanisms that set performance and pedestal profiles is being advanced through gyrokinetic simulations, reduced model development, comparison NSTX experiment, as well improved simulation RF heating. develop stable non-inductive scenarios needed for steady-state operation,...
Abstract The spherical tokamak (ST) is a leading candidate for Fusion Nuclear Science Facility (FNSF) due to its compact size and modular configuration. National Spherical Torus eXperiment (NSTX) MA-class ST facility in the US actively developing physics basis an ST-based FNSF. In plasma transport research, experiments exhibit strong (nearly inverse) scaling of normalized confinement with collisionality, if this trend holds at low high fusion neutron fluences could be achievable very...
National Spherical Torus Experiment [which M. Ono et al., Nucl. Fusion 40, 557 (2000)] high-power divertor plasma experiments have shown, for the first time, that benefits from lithium coatings applied to facing components found previously in limited plasmas can occur also diverted configurations. Lithium were with pellets injected into helium discharges, and an oven directed a collimated stream of vapor toward graphite tiles lower center stack divertor. depositions few milligrams 1g been...
Abstract A fusion nuclear science facility (FNSF) could play an important role in the development of energy by providing environment needed to develop materials and components. The spherical torus/tokamak (ST) is a leading candidate for FNSF due its potentially high neutron wall loading modular configuration. key consideration choice configuration range achievable missions as function device size. Possible include: fluence, demonstrating tritium self-sufficiency, electrical self-sufficiency....
Abstract Lithium wall coatings have been shown to reduce recycling, suppress edge-localized modes (ELMs), and improve energy confinement in the National Spherical Torus Experiment (NSTX). Here we document effect of gradually increasing lithium on discharge characteristics, with reference ELMy discharges obtained boronized, i.e. non-lithiated conditions. We observed a continuous but not quite monotonic reduction recycling improvement confinement, gradual alteration edge plasma profiles,...
Extensive lithium wall coatings and liquid plasma-limiting surfaces reduce recycling, with dramatic improvements in Ohmic plasma discharges the Current Drive Experiment-Upgrade. Global energy confinement times increase by up to 6 times. These results exceed scalings such as $\mathrm{ITER}98\mathrm{P}(y,1)$ $2--3$ times, represent largest ever observed for an tokamak plasma. Measurements of ${D}_{\ensuremath{\alpha}}$ emission indicate that global recycling coefficients decrease approximately...
Liquid metal plasma-facing components (PFCs) have been proposed as a means of solving several problems facing the creation economically viable fusion power reactors. To date, few demonstrations exist this approach in diverted tokamak and we here provide an overview such work on National Spherical Torus Experiment (NSTX). The Lithium Divertor (LLD) was installed operated for 2010 run campaign using evaporated coatings filling method. LLD consisted copper-backed structure with porous...
Lithium wall coatings have been shown to reduce recycling, improve energy confinement, and suppress edge localized modes in the National Spherical Torus Experiment. Here, we show that these effects depend continuously on amount of predischarge lithium evaporation. We observed a nearly monotonic reduction decrease electron transport, modification profiles stability with increasing lithium. These correlations challenge basic expectations, given even smallest exceeded needed for nominal...
The snowflake magnetic configuration is characterized by the presence of two closely spaced poloidal field nulls that create a characteristic hexagonal (reminiscent snowflake) separatrix structure. properties and plasma behaviour in are determined simultaneous action both nulls, this generating lot interesting physics, as well providing chance for improving divertor performance. Among potential beneficial effects geometry are: increased volume low around null, connection length, heat flux...
We report the first observation of global recycling coefficient R near 0.5 in Lithium Tokamak eXperiment-β (LTX-β), significantly below minimum previously reported other devices. In a series experiments with varied Li wall conditioning, estimates have been made using Lyman-α array and DEGAS2 modeling. A progressive reduction emission increased lithium an increase edge electron temperature are observed. It is also observed that increasing coating thickness, effective particle confinement time...
Recent experiments in the Current Drive Experiment-Upgrade (CDX-U) provide a first-ever test of large area liquid lithium surfaces as tokamak first wall to gain engineering experience with metal and investigate whether very low recycling plasma regimes can be accessed walls. The CDX-U is compact (R = 34 cm, 22 Btoroidal 2 kG, IP 100 kA, Te(0)∼ eV, ne(0) ∼ 5 × 1019 m−3) spherical torus at Princeton Plasma Physics Laboratory. A toroidal pool limiter an 2000 cm2 (half total limiting surface)...
Steady-state handling of divertor heat flux is a critical issue for ITER and future conventional spherical tokamaks with compact high-power density divertors. A novel ‘snowflake’ (SFD) configuration was theoretically predicted to have significant magnetic geometry benefits mitigation, such as an increased plasma-wetted area higher volume available volumetric power momentum loss processes, compared the standard divertor. Both peak reduction impurity screening been achieved simultaneously core...
Experimental results from NSTX indicate that the snowflake divertor (D. Ryutov, Phys. Plasmas 14, 064502 (2007)) may be a viable solution for outstanding tokamak plasma-material interface issues. Steady-state handling of heat flux and plate erosion remains to critical issues ITER future concept devices based on conventional spherical geometry with high power density divertors. Experiments conducted in 4–6 MW NBI-heated H-mode plasmas demonstrated is compatible high-confinement core plasma...
The coating of plasma facing components (PFCs) with lithium improves energy confinement and eliminates ELMs in the National Spherical Torus Experiment, latter due to a relaxation density pressure profiles that reduces drive for peeling-ballooning modes. 2-D interpretive transport modeling discharges without shows reduction PFC recycling coefficient from R ∼ 0.98 0.90 is required match drop Dα emission coatings. A broadening edge barrier region showing reduced coefficients observed, ∼75% D χe...
The National Spherical Torus Experiment (NSTX) has undergone a major upgrade, and the NSTX Upgrade (NSTX-U) Project was completed in summer of 2015.NSTX-U first plasma subsequently achieved, diagnostic control systems have been commissioned, H-mode accessed, magnetic error fields identified mitigated, physics research campaign carried out.During ten run weeks operation, NSTX-U surpassed record pulse-durations toroidal (TF), high-performance ~1 MA plasmas comparable to best sustained near...
Transport and turbulence profiles were directly evaluated using probes for the first time in edge scrape-off layer (SOL) of NSTX [Ono et al., Nucl. Fusion 40, 557 (2000)] low (L) high (H) confinement, power (Pin∼ 1.3 MW), beam-heated, lower single-null discharges. Radial turbulent particle fluxes peak near last closed flux surface (LCFS) at ≈4×1021 s−1 L-mode are suppressed to ≈0.2×1021 H mode (80%–90% lower) mostly due a reduction density fluctuation amplitude phase between radial velocity...
H-mode operation plays a crucial role in National Spherical Torus Experiment (NSTX) research, allowing higher beta limits due to reduced plasma pressure peaking, and long pulse high bootstrap current fraction. Here, new results are presented the areas of edge localized modes (ELMs), pedestal physics power threshold studies. ELMs several types as reported by aspect ratio tokamaks have been observed: (1) large, Type I ELMs, (2) intermediate-sized III (3) tiny ELMs. Many performance discharges...
The major objective of the National Spherical Torus Experiment (NSTX) is to understand basic toroidal confinement physics at low aspect ratio and high βT in order advance spherical torus (ST) concept. In do this, NSTX utilizes up 7.5 MW neutral beam injection, 6 harmonic fast waves (HHFWs), it operates with plasma currents 1.5 MA elongations 2.6 a field 0.45 T. New facility, diagnostic modelling capabilities developed over past two years have enabled research team make significant progress...